The following relates to the nuclear reactor arts, electrical power generation arts, nuclear safety arts, and related arts.
Nuclear reactors employ a reactor core comprising a mass of fissile material, such as a material containing uranium oxide (UO2) that is enriched in the fissile 235U isotope. Primary coolant water, such as light water (H2O) or heavy water (D2O) or some mixture thereof, flows through the reactor core to extract heat for use in heating secondary coolant water to generate steam or for some other useful purpose. For electrical power generation, the steam is used to drive a generator turbine. In thermal nuclear reactors, the primary coolant water also serves as a neutron moderator that thermalizes neutrons, which enhances reactivity of the fissile material. Various reactivity control mechanisms, such as mechanically operated control rods, chemical treatment of the primary coolant with a soluble neutron poison, or so forth are employed to regulate the reactivity and resultant heat generation. In a pressurized water reactor (PWR), the primary coolant water is maintained in a subcooled state in a sealed pressure vessel that also contains the reactor core, and the liquid primary coolant water flows through a steam generator located outside the pressure vessel or inside the pressure vessel (the latter being known as an integral PWR) to generate steam to drive a turbine. In a boiling water reactor (BWR), the primary coolant boils in the pressure vessel and is piped directly to the turbine. Some illustrative examples of integral PWR designs are set forth in Thome et al., “Integral Helical Coil Pressurized Water Nuclear Reactor”, U.S. Pub. No. 2010/0316181 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety, and in Malloy et al., “Compact Nuclear Reactor”, U.S. Pub. No. 2012/0076254 A1 published Mar. 29, 2012 which is incorporated herein by reference in its entirety. These are merely illustrative examples.
In either a PWR or a BWR, the primary coolant water is maintained at controlled elevated temperature and pressure by heat generated in the radioactive nuclear reactor core balanced by heat sinking provided by steam generation and subsequent condensation (i.e. a steam cycle). In the event of a reactor vessel breach (known in the art as a loss of coolant accident, i.e. LOCA), primary coolant flashes to steam outside the pressure vessel. A radiological containment (sometimes called primary containment or simply containment) surrounds the pressure vessel to contain any such steam release, and an automatic reactor shutdown is performed to extinguish the nuclear reaction, typically including scram of control rods and optionally injection of borated water or another soluble neutron poison into the primary coolant in the pressure vessel. An emergency core cooling system (ECCS) and/or other safety systems also respond by removing decay heat from the nuclear reactor, condensing and recapturing any primary coolant steam released into the radiological containment, and depressurizing the reactor pressure vessel.
The pressure vessel depressurization entails venting primary coolant, typically in the form of steam, to a dedicated compartment, and/or into the containment or other sink. Initially, the high pressure inside the vessel provides substantial mass transport for the venting. As the pressure decreases, the mass transport rate for a given vent orifice decreases. To vent to atmospheric pressure in a reasonable time frame, the venting system opens additional valves as the pressure in the vessel decreases to increase the total orifice area and maintain a reasonable mass transport rate. Redundant valving must be provided in accord with safety regulations of the United States Nuclear Regulatory Commission (U.S. NRC; similar regulations apply in most other countries). This increases the cost and complexity of the venting system; and, even with redundant valving, there is the potential for failure at multiple points, including at the valve actuator or at electronics producing the actuation signal. Such a failure can delay reactor shutdown and increase the time the reactor remains pressurized in an abnormal state.
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